ASTM C1562-2010 Standard Guide for Evaluation of Materials Used in Extended Service of Interim Spent Nuclear Fuel Dry Storage Systems《临时废核燃料干燥储存系统长期运行过程中使用的材料的评估的标准指南》.pdf

《ASTM C1562-2010 Standard Guide for Evaluation of Materials Used in Extended Service of Interim Spent Nuclear Fuel Dry Storage Systems《临时废核燃料干燥储存系统长期运行过程中使用的材料的评估的标准指南》.pdf》由会员分享，可在线阅读，更多相关《ASTM C1562-2010 Standard Guide for Evaluation of Materials Used in Extended Service of Interim Spent Nuclear Fuel Dry Storage Systems《临时废核燃料干燥储存系统长期运行过程中使用的材料的评估的标准指南》.pdf（25页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: C1562 10Standard Guide forEvaluation of Materials Used in Extended Service of InterimSpent Nuclear Fuel Dry Storage Systems1This standard is issued under the fixed designation C1562; the number immediately following the designation indicates the year oforiginal adoption or, in the case

2、of revision, the year of last revision. A number in parentheses indicates the year of last reapproval. Asuperscript epsilon () indicates an editorial change since the last revision or reapproval.1. Scope1.1 Part of the total inventory of commercial spent nuclearfuel (SNF) is stored in dry cask stora

3、ge systems (DCSS) underlicenses granted by the U.S. Nuclear Regulatory Commission(NRC). The purpose of this guide is to provide information toassist in supporting the renewal of these licenses, safely andwithout removal of the SNF from its licensed confinement, forperiods beyond those governed by th

4、e term of the originallicense. This guide provides information on materials behaviorunder conditions that may be important to safety evaluationsfor the extended service of the renewal period. This guide iswritten for DCSS containing light water reactor (LWR) fuelthat is clad in zirconium alloy mater

5、ial and stored in accor-dance with the Code of Federal Regulations (CFR), at anindependent spent-fuel storage installation (ISFSI).2The com-ponents of an ISFSI, addressed in this document, include thecommercial SNF, canister, cask, and all parts of the storageinstallation including the ISFSI pad. Th

6、e language of this guideis based, in part, on the requirements for a dry SNF storagelicense that is granted, by the U.S. Nuclear Regulatory Com-mission (NRC), for up to 20 years. Although governmentregulations may differ for various nations, the guidance onmaterials properties and behavior given her

7、e is expected tohave broad applicability.1.2 This guide addresses many of the factors affecting thetime-dependent behavior of materials under ISFSI service 10CFR Part 72.42. These factors are those regarded to beimportant to performance, in license extension, beyond thecurrently licensed 20-year per

8、iod. Examples of these factorsare given in this guide and they include materials alterations orenvironmental conditions for components of an ISFSI systemthat, over time, could have significance related to safety. Forpurposes of this guide, a license period of an additional 20 to80 years is assumed.1

9、.3 This guide addresses the determination of the conditionsof the spent fuel and storage cask materials at the end of theinitial 20-year license period as the result of normal events andconditions. However, the guide also addresses the analysis ofpotential spent fuel and cask materials degradation a

10、s the resultof off-normal, and accident-level events and conditions thatmay occur during any period.1.4 This guide provides information on materials behaviorto support continuing compliance with the safety criteria,which are part of the regulatory basis, for licensed storage ofSNF at an ISFSI. The s

11、afety functions addressed and discussedin this standard guide include thermal performance, radiologi-cal protection, confinement, sub-criticality, and retrievability.The regulatory basis includes 10 CFR Part 72 and supportingregulatory guides of the U.S. Nuclear Regulatory Commission.The requirement

12、s set forth in these documents indicate that thefollowing items were considered in the original licensingdecisions: properties of materials, design considerations fornormal and off-normal service, operational and natural events,and the bases for the original calculations. These items mayrequire reco

13、nsideration of the safety-related arguments thatdemonstrate how the systems continue to satisfy the regulatoryrequirements. Further, to ensure continued safe operation, theperformance of materials must be justified in relation to theeffects of time, temperature, radiation field, and environmentalcon

14、ditions of normal and off-normal service. Arguments forlong-term performance must account for materials alterations(especially degradations) that are expected during the serviceperiods, which include the periods of the initial license and ofthe license renewal. This guide pertains only to structures

15、,systems, and components important to safety during extendedstorage period and during retrieval functions, including trans-port and transfer operations. Materials information that per-tains to safety functions, including retrieval functions, ispertinent to current regulations and to license renewal

16、process,1This guide is under the jurisdiction of ASTM Committee C26 on Nuclear FuelCycle and is the direct responsibility of Subcommittee C26.13 on Spent Fuel andHigh Level Waste.Current edition approved Jan. 1, 2010. Published March 2010. Originallyapproved in 2003. Last previous edition approved i

17、n 2003 as C1562 032. DOI:10.1520/C1562-10.2In general fuels of higher burnup (45 MWd/kgU) and MOX fuels are notincluded in this guide. Guidance for these fuels are expected to be included in anAnnex to be written later.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohoc

18、ken, PA 19428-2959, United States.and this information is the focus of the guide. This guide is notintended to supplant the existing regulatory process.2. Referenced Documents2.1 ASTM Standards:3C33 Specification for Concrete AggregatesC227 Test Method for Potential Alkali Reactivity ofCement-Aggreg

19、ate Combinations (Mortar-Bar Method)C295 Guide for Petrographic Examination of Aggregatesfor ConcreteC859 Terminology Relating to Nuclear MaterialsC1174 Practice for Prediction of the Long-Term Behaviorof Materials, Including Waste Forms, Used in EngineeredBarrier Systems (EBS) for Geological Dispos

20、al of High-Level Radioactive Waste2.2 Government Documents:410 CFR Part 50 Domestic Licensing of Production andUtilization Facilities10 CFR Part 60 Disposal of High Level Radioactive Wastesin Geologic Repositories10 CFR Part 63 Disposal of High Level Radioactive Wastesin a Proposed Geologic Reposito

21、ry in Yucca Mountain10 CFR Part 71 Packaging and Transport of RadioactiveMaterials10 CFR Part 72 Licensing Requirements for the Indepen-dent Storage of Spent Nuclear Fuel and High-LevelRadioactive Waste2.3 NUREG Standards:5NUREG-1536 Standard Review Plan for Dry Storage CaskSystems, January 1997NURE

22、G-1567 Standard Review Plan for Spent Fuel DryStorage Facilities, Report, January 1998NUREG-1571 Information Handbook on IndependentSpent Fuel Storage InstallationsNUREG/CR-6407 Classification of Transportation Packag-ing and Dry Spent Fuel Storage System ComponentsAccording to Importance to Safety,

23、 February, 1996, INELReport 95/0551ISG-1 Interim Staff Guidance Number 1, U.S. NRC, SpentFuel Project Office2.4 American Concrete Institute Standards:6ACI 201.2R-97 Guide to Durable ConcreteACI 209R-97 Prediction of Creep, Shrinkage and Tempera-ture Effects in Concrete StructuresACI 301-99 Building

24、Code Requirements for ReinforcedConcreteACI 318-02 Building Code Requirements for ReinforcedConcreteACI 349-00 Code Requirements for Nuclear Safety RelatedConcrete StructuresACI 359-01 Code for Concrete Reactor Vessels and Con-tainments, also designated as ASME Boiler and PressureVessel Code, Sectio

25、n III, Div 2, Code for ConcreteReactor Vessels and Containments2.5 ANSI Documents:7ANSI/ANS-6.4-1985 Guidelines on the Nuclear Analysisand Design of Concrete Radiation Shielding for NuclearPower PlantsANSI/ANS-57.9 Design Criteria for an Independent SpentFuel Storage Installation (Dry Storage Type)A

26、NSI/ANS-57.10 Design Criteria for Consolidation ofLWR Spent Fuel2.6 Other Documents:ASME-B Revision 1, TR-103842, July 199493. Terminology3.1 The terminology of Terminology C859 applies to thisdocument except as given below.3.2 Definitions of Terms Specific to This Standard:3.2.1 accident-level even

27、ts or conditionsthe extreme levelof an event or condition for which there is a specifiedresistance, limit of response, and requirement for a given levelof continuing capability, which exceed “off-normal” events orconditions. They include both design basis accidents anddesign-basis for natural phenom

28、ena events and conditions.NUREG-1536, NUREG-1567NOTE 1Specific accident conditions to be addressed have beenevaluated for each dry cask storage system (DCSS) and are documentedin a Safety Analysis Report for that system.3.2.2 alteration modea particular form of alteration, forexample, general corros

29、ion, passivation. C11743.2.3 ASTM guidea compendium of information or seriesof options that does not recommend a specific course of action.3.2.4 canisterin a dry cask storage system (DCSS) forspent nuclear fuel, a metal cylinder that is sealed at both endsand is used to perform the function of confi

30、nement, while aseparate overpack performs the functions of shielding andprotection of the canister from the effects of impact loading.3.2.5 caskin a dry cask storage system (DCSS) for spentnuclear fuel, a stand-alone device that performs the functionsof confinement, radiological shielding, and physi

31、cal protectionof spent fuel during normal, off-normal, and accident condi-tions. NUREG-15713.2.6 certificate of compliancein a dry cask storagesystem (DCSS) for spent nuclear fuel, a certificate issued by theU.S. Nuclear Regulatory Commission (NRC) to the designer/3For referenced ASTM standards, vis

32、it the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.4Available from U.S. Government Printing Office Superintendent of Documents,732 N. Capitol St.,

33、NW, Mail Stop: SDE, Washington, DC 20401, http:/www.access.gpo.gov.5Available from National Technical Information Service (NTIS), 5285 PortRoyal Rd., Springfield, VA 22161, http:/www.ntis.gov.6Available from American Concrete Institute (ACI), P.O. Box 9094, FarmingtonHills, MI 48333-9094, http:/www.

34、aci-int.org.7Available from American National Standards Institute (ANSI), 25 W. 43rd St.,4th Floor, New York, NY 10036, http:/www.ansi.org.8Available from American Society of Mechanical Engineers (ASME), ASMEInternational Headquarters, Three Park Ave., New York, NY 10016-5990, http:/www.asme.org.9Av

35、ailable from Electric Power Research Institute (EPRI), 3512 Hillview Ave.,Palo Alto, CA 943041395.C1562 102vendor of a specific cask model that meets the requirements setforth in 10 CFR Part 72.236.3.2.7 confinementin a dry cask storage system (DCSS) forspent nuclear fuel, the ability to prevent the

36、 release ofradioactive substances into the environment. NUREG-15713.2.8 confinement systemsin a dry cask storage system(DCSS) for spent nuclear fuel, the assembly of components ofthe packaging intended to retain the radioactive material duringstorage. These may include the cladding, storage system s

37、hell,bottom and lid, penetration covers, the closure welds or sealsand bolts and other components. NUREG-15363.2.9 criticalityin a dry cask storage system (DCSS) forspent nuclear fuel, the condition wherein a system or mediumis capable of sustaining a nuclear chain reaction. C8593.2.10 degradationan

38、y change in the properties of amaterial that adversely affects the behavior of that material;adverse alteration. C11743.2.11 degraded claddingin spent nuclear fuel, claddingmaterial that by visual inspection appears to be structurallydeformed or damaged to an extent that special handling isexpected

39、to be required. See ISG-1 for damaged fuel.3.2.12 dry cask storage system (DCSS)in nuclear wastemanagement, a set of components that performs the functionsof confinement, radiological shielding, and physical protectionof spent nuclear fuel during normal, off-normal, and accidentconditions. Examples

40、would include canister-based systemswith their metal or concrete overpack or vault, or an integratedcask.3.2.13 dry storagein nuclear waste management, thestorage of spent nuclear fuel after removal of the water fromthe fuel, cladding and all components of a dry cask storagesystem, and after the atm

41、osphere has been replaced with aninert atmosphere.3.2.14 independent spent fuel storage installation (ISFSI)any complex designed and constructed for interim dry storageof spent nuclear fuel and other radioactive materials associatedwith spent fuel storage. It must meet the requirements in 10CFR Part

42、 72. In this guide a Monitored Retrievable Storage(MRS) site is also considered an ISFSI. NUREG-15713.2.15 monitoringin a dry cask storage system (DCSS)for spent nuclear fuel, testing and data collection to determinethe status of a DCSS and to verify the continued efficacy of thesystem, on the basis

43、 of measurements of specified parametersincluding temperature, radiation, functionality and/or charac-teristics of components of the system.3.2.16 normal events and conditionsthe maximum levelof an event or condition expected to routinely occur.NOTE 2Specific normal conditions to be addressed have b

44、een evalu-ated for each licensed DCSS and are documented in a Safety AnalysisReport for that system.3.2.17 off-normal events or conditionsin a dry cask stor-age system (DCSS) for spent nuclear fuel, the maximum levelof an event that, although not occurring regularly, can beexpected to occur with mod

45、erate frequency, and for whichthere is a corresponding maximum specified resistance, limit ofresponse, or requirement for a given level of continuingcapability. NUREG-1536NOTE 3Specific off-normal conditions to be addressed have beenevaluated for each licensed DCSS and are documented in a SafetyAnal

46、ysis Report for that system.3.2.18 radiation shieldingin a dry cask storage system(DCSS) for spent nuclear fuel, barriers to radiation, which aredesigned to meet the requirements of 10 CFR Parts 72.104(a),and 72.106(b), and 72.128(a.2).3.2.19 retrievabilityin a dry cask storage system (DCSS)for spen

47、t nuclear fuel, the ability to remove spent nuclear fuelfrom storage for further processing or disposal. 10 CFRPart 72.122 (1)3.2.20 safety analysis report (SAR)in a dry cask storagesystem (DCSS) for spent nuclear fuel, the document that issupplied by a DCSS vendor or site-specific independent spent

48、fuel storage installation (ISFSI) applicant to the U.S. NuclearRegulatory Commission (NRC) for analysis and confirmingcalculations (review and approval). NUREG-15713.2.21 safety evaluation report (SER)in a dry cask stor-age system (DCSS) for spent nuclear fuel, the document thatthe U.S. Nuclear Regu

49、latory Commission (NRC) publishesafter review of a Safety Analysis Report (SAR). NUREG-15713.2.22 service conditionsin a dry cask storage system(DCSS) for spent nuclear fuel, the time of service, tempera-tures, environmental conditions, radiation, and loading, etc.that a component experiences during storage.3.2.23 spent nuclear fuel (SNF), spent fuelnuclear fuelthat has undergone at least one year of decay since being usedas a source of energy in a power reactor, and has not beenseparated into its constituent elements by reprocessing.NUREG-1571NOTE